ML20217H417
| ML20217H417 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/06/1997 |
| From: | Gundrum L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217H421 | List: |
| References | |
| NUDOCS 9708110143 | |
| Download: ML20217H417 (10) | |
Text
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NUCLEAR REEULATCRY COMMISSION I
2 WASHINGTON. D.C. 30seHo01 4,.....,o WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.175 License No. DPR-24 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated August 22. 1996, as supplemented on July 14 1997. complies with the standards and re Energy Act of 1954, as amended (the Act)quirements of the Atomic
, and the Comission's rules and regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the application the provisions of the Act, and the rules and regulations of the Commission:
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9708110143 970806 PDR ADOCK 05000266 P
2-2.
Accordingly, paragraph 3.A of Facility Operating License No. DPR-24 is hereby amended to read as follows:
A.
Maximum Power Levels-The licensee is authorized to o~perate the facility at reactor core power levels not in excess of 1518.5 megawatts thermal.
3.
This license amendment is-effective imediately upon issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Nash b nchtano Linda L. Gundrum. Proiect Manager Project Directorate I I-1 Division of Reactor Projects - III/IV -
Office of Nuclear Reactor Regulation
-Attachments:
1.
Page 3 of the License
- 2.
Revised Bases pages 15.3.1-7 and 15.3.1-8 Date of issuance:
August 6, 1997
- Page 3 of the license is attached, for convenience, for the composite license to reflect this change.
o
I l
Wisconsin Electric Power Company _ I 3.
This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations:
10 CFR Part 20. Section 30.34 of 10 CFR Part 30. Section 40.41 of 10-CFR Par't 40. Sections 50.54 and 50.59 of 10 CFR Part'50, and Section 70.32 of 10 CFR Part 70: and is subject to all_ applicable provisions of the Act and to the rules, regulations, and orders of the Comission now or hereafter in effect: and is subject to the additional conditions specified below:
A.
Maximum Power Levels The licensee is authorized to operate the facility at reactor core power levels not in excess of 1518.5 megawatts thermal.
l B.
Technical Soecifications The Technical Specif Nations contained in Appendices A and B.
as revised through Amendment No.
174, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with Technical Specifications.
-C.
Reoort The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.
D.
Records
.The licensee shall keep facility ' operating records in accordance with the requirements of the Technical Specifications.
E.
Soent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its-storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended.
In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on_ every poison assembly shall be performed.
Point Beach Unit 1 Amendment No.175 l
J
of the vessel is computed to be 2.5 x 10" neutrons /cm for 40 years of 2
operation at 1518.5 MWt and 80 percent load factor.* This maximum fluence is l the exposure expected at the inner reactor vessel wall.
However, the neutron fluence used to predict the ARTa shift is the one-quarter shell thickness neutron exposure. The relationship between fluence at the vessel ID wall and the fluence at the one quarter and three-quarter shell thickness locations is as presented in Regulatory Guide 1.99 Revision 2. " Radiation Damage to Reactor Vessel Materials." (Reference 6)
Once the fluence is determined, the adjusted reference temperature used in revising the heatup and cooldown curves is obtained by utilizing the method in Section 1.1 of Regulatory Guide 1.99 Revision 2 (Reference 6) for the limiting weld material of both Unit 1 and Unit 2.
The heatup and cooldown curves presented in Figure 15.3.1-1 and 15.3.1-2 were calculated based on the above information and the methods of ASME Code Section III (1974 Edition). Appendix G. " Protection Against Nonductile Failure", and are applicable up to the ope':tional exposure indicated on the
- figures, The regulations governing the pressure-temperature limits (10 CFR 50 -
Appendix G and ASME Code Section III - Appendix G) do not require additional margins for instrumentation uncertainties be added to the heatup and cooldown curves. This is because the inclusion of instrumentation uncertainties, in addition to other conservatisms in the methods for calculating the pressure temperature limits. is not necessary to protect the vessel from damage.
Unit 1 - Amendment No.
93.125.158, 175 Unit 2 - Amer.dment No. 102.129.172, 179 15.3.1-7 i
The actual temperature shift of the vessel material will be established periodically during operation by removing ano evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius are identified by a specified lead Mtor. the measured temperature shift for a sample is an excellent P
Aor of the effects of power operation on the adjacent section of the
.. tor vessel.
If the experimental temperature shift (at the 30 ft-lb level) does not substantiate the predicted shift, new prediction curves and heatup and cooldown curves must be developed.
The pressure-temperature limit lines shown on Figure 15.3.1-1 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.
The spray should not be used if the temperature difference between the pressurizer and spray fluid is greater than 320 F.
This limit is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.
The temperature requirements for the steam generator correspond with the measured NDT for the shell.
The reactor vessel materials surveillance capsule removal schedules have been developed based upon the requirements of the Code of Federal Regulation, Title
- 10. Part 50. Appendix H. and with consideration of ASTM Standard E-185-82.
When the capsule lead factors are considered, the scheduled removal dates accommodate the weld data needs of all the participants in the Babcock and Wilcox Master integrated Reactor Vessel Surveillance Program.
Additionally, the schedule will provide plate / forging material data as well as fluence data corresponding to the expiration of the current licenses and of any future license extensions.
References (1)
FSAR, Section 4.1.5 (2)
Westinghouse Electric Corporation. WCAP-12794. Rev. 3/12795, Rev. 3 l
(3)
Westinghouse Electric Corporation. WCAP-8743 (4)
Westinghouse Electric Corporation WCAP-8738 (5)
Babcock & Wilcox. BAW 1803 (6)
Regulatory Guide 1.99. Revision 2 Unit 1 - Amendment No. -125.131.158,175 Unit 2 - Amendment No. 129.135.172. 179 15.3.1-8
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NUCLEAR RESULAT@RY C2MMISSION i
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WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.179 License No. DPR-27 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated August 22. 1996, as supplemented on July 14, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954. as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the application, the arovisions of the Act, and the rules and regulations of the Commission:
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public:
and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
4
. 2.
Accordingly, paragraph 3.A of Facility Operating License No. DPR-27 is hereby amended to read as follows:
A.
Maximum Power Levels The licensee is authorized to operate the facility at reactor' core power _ levels not in excess of 1518.5 megawatts thermal.
3.
This license amendment is effective immediately upon issuance.
FOR THE NUCLEAR REGULATORY COMMISSION d6v hbdssmo LindaL.Gundrum.ProjectManager Project Directorate 111-1 Division of Fnactor Projects - III/IV Office of Nuclear Reactor Regulation Attachments:
1.
Page 3 of the License" 2.
Revised Bases pages 15.3.1-7 and 15.3.1-8
'Date of issuance:
August 6, 1997 I
i "Page 3 of the license is attached. for convenience, for the composite license to reflect this change.
.J
4 -
Wisconsin Electric Power Company.
3.
This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations:
10 CFR Part 20. Section 30.34 of 10 CFR Part 3D. Section 40.41 of 10 CFR Part 40. Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70: andissubjecttoallapplicable provisions of the Act and to the rules, regulations, and orders of the Comission now or hereafter in effect; and is subject to the additional conditions specified below:
A.
Maximum Power Levels The licensee is authorized to operate the facility at. reactor core power levels not in excess of 1518.5 megawatts thermal.
l B.
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
178, are hereby incorporated in the license. The licensee shall operate the facility in accordance with Technical Specifications.
C.
Reoort The licensee shall make certain reports in accordance with t' a requirements of the Technical Specifications.
D.
Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.
E.
Soent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to' increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended.
In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.
-Point Ba ch Unit 2 Amendment No.179
) _f
i of the vessel-is computed to be 2.5 x 10" neutrons /cm for 40 years of r
. operation at 1518.5 MWt and 80 percent load factor This maximum fluence is l the exposure expected at the inner reactor vessel wall. However, the r.eutron i
fluence used to predict the ART,m shift is the one quarter shell thickness neutron exposure. The relationehip between fluence at the vessel 10 wall and the fluence at the one-quarter and three-quarter shell thickness locatims is as presented in Regulatory Guide 1.99 Revision 2. " Radiation Damage to Reactor Vessel Materials." (Reference 6)
Once the fluence is determined, the adjusted reference temperature used in revising the heatup and cooldown curves is obtained by utilizing the method in Section 11 of Regulatory Guide 1.99 Revision 2 (Reference 6) for the limiting weld material of both Unit 1 and Unit 2.
The heatup and c aldown curves presented in Figure 15.3.1-1 and 15.3.1-2 were calculated based on the above information and the methods of ASME Code Section III (1974 Edition) Appendix G. " Protection Against Nonductile Failure", and are applicable up to the operational exposure indicated on the figures.
The regulations governing the pressure-temperature limits (10 CFR 50 -
-Appendix G and ASME Code Section III - Appendix G) do not require additional margins for instrumentation uncertainties be added to the heatup and cooldown curves. This is because the inclusion of instrumentation uncertainties, in addition to other conservatisms in the methods for calculating the pressure.
temperature limits, is not necessary to protect the vessel from damage.
Unit 1 - Amendment No.
98.125.158,175 Unit 2 - Amendment No. 102.129.172,179 15.3.1-7 j
The actual temperature shift of the vessel material will be established pericdically during operation by reinoving and evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius are identified by a specified lead factor, the measured temperature shift for a sample is an excellent d
E indicator of the effects of power operation on the adjacent section of the reactor vessel.
If the experimental temperature shift (at the 30 ft-lb level) does not substantiate the predicted shift, new prediction curves and heatup and cooldown curves must be developed.
The pressure temperature limit lines shown on Figure 15.3.1-1 for reactor criticality and for inservice leak and hydrostati: testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.
The spray sh' 'i not be used if the temperature difference between the pressurizer c spray fluid is greeter than 320 F.
This limit is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.
The temperature requirements for the steam generator correspond with the measured NDT for the shell.
The reactor vessel materials surveillance capsule removal schedules have been developed based upon the requirements of the Code of Federal Regulation. Title
- 10. Part 50. Appendix H. and with consideration of ASTM Standard E-185-82.
When the capsule lead factors are considered, the scheduled removal dates accommodate the weld data needs of all the participants in the Babcock and Wilcox Mast 2r integrated Reactor Vessel Surveillance Program. Additionally, the schedule will provide plate / forging material data as well as fluence data corresponding to the expiration of the current licenses and of any future license extensions.
M erences (1)
FSAR. Section 4.1.5 (2)
Westinghouse Electric Corporation. WCAP-12794. Rev. 3/12795 Rev. 3 l
(3)
Westinghouse Electric Corporation. WCAP-8743 (4)
Westinghouse Electric Corporatien. WCAP-8738 (5)
Babcock & Wilcox BAW 1803 (6)
Regulatory Guide 1.99 Revision 2 Unit 1 - Amendment No. 125.131.168.175 Unit 2 - Amendment No. 129.135.172.179 15.3.1-8
_